DOCSLIB.ORG

Overview of the KSTAR Research Progress and Future Plan Toward ITER and K-DEMO

Total Page：16

File Type：pdf, Size：1020Kb

Download full-text PDF Read full-text
Overview of the KSTAR Research Progress and Future Plan Toward ITER and K-DEMO Overview of the KSTAR Research Progress and Future Plan toward ITER and K-DEMO a Ulsan National Institute of Science and Technology, Ulsan, b b a b Hyeon K. Park M.J. CHOI , S.H. HONG Y. In , Y.M. JEON , Korea J.S. Kob, W.H. Kob, J.G. KWAKb J.M. KWONb, J. b National Fusion Research Institute, Daejeon, Korea UNIST, Ulsan, Korea Leeb, J.H. LEEb, W. LEEb, Y.B. NAMd, Y.K. Ohe, c Seoul National University, Seoul, Korea b f g d CEA, St. Paul Lez Durance, France B.H. PARK , J.K. PARK , Y.S. PARK , S.J. e b c b h ITER Organization, Route de Vinon-sur Verdon. 13067 St. WANG , M. YOO , S.W. YOON , J.W. AHN , Paul Lez Durance, France On behalf of b f i b J.G. BAK , C.S. CHANG , W.H. CHOE , Y. CHU , f Princeton Plasma Physics Laboratory, Princeton, NJ, USA the co-authors and KSTAR Team J. CHUNGb, N. EIDIETISj, H.S. HANb, S.H. g Columbia University, NY, USA HAHNb, H.G. JHANGb, J.W. JUHNb, J.H. KIMb, h Oak Ridge National Laboratory, TN, USA i K. KIMh, A. LOARTEe, H.H. LEEb, K.C. LEEb, D. Korea Advanced Institute of Science and Technology, Daejeon, Korea th f c b b at 27 IAEA FEC, MUELLER , Y.S. Na , Y.U. NAM , G.Y. PARK , j General Atomics, CA, USA b e g K.R. PARK , R.A. PITTS , S.A. SABBAGH , k Pohang University of Science and Technology, Pohang, Gandhinagar, India, 2018 G.S. YUNk Korea IAEA FEC PARK - 1 - Topics to be addressed . Mission and role of the KSTAR research for ITER and K- DEMO . Advantages of tokamak plasma research on KSTAR . Progress of the KSTAR research program . High performance and steady-state operation . Physics of MHD/Turbulence . Progress and understanding of the ELM-crash control with RMP . Research efforts for ITER and K-DEMO . Future upgrade plan . Summary IAEA FEC PARK - 2 - Mission and role of KSTAR for ITER and K-DEMO KSTAR Mission is: • to explore science and technology for high performance steady-state operation, that are essential for fusion reactors K-DEMO Role of KSTAR for ITER & K-DEMO • Expansion of operating regimes, validation of tokamak physics and advanced control for SC device like ITER • Stationary High beta (흱N ~4) with high using the KSTAR unique tools bootstrap • Optional new mode of operation including to expand the KSTAR operation boundaries of the conventional modes of auto start-up with high duty cycle operation and new mode of operation for minimum harmful instabilities • CD and divertor heat flux control (avoid NTM and ELM-crash) to explore physics and preemptive control of the harmful instabilities in high beta operation (ELM-crash, NTM, disruptions, etc.) ITER to assist optimum operation of ITER: H-mode threshold power control, rotation control via NTV, heat dispersion with RMP, etc. • Exploration of the limit of high beta steady state operation and R&D of key engineering and technology for K-DEMO • Steady-state H-mode (흱N ~2) Exploration of higher beta operation with the KSTAR upgrade • Rotation by NTV, and new start-up, KSTAR H-mode access Alternative current drive for steady-state operation • Suppression of ELM-crash, New start-up for high duty cycle with long pulse disruption IAEA FEC PARK - 3 - View of KSTAR with new NBI-2 system ECH & Helicon & LHCD KSTAR NBI-1 Advanced Microwave Imaging Diagnostics Carbon wall NBI-2 (New) IAEA FEC PARK - 4 - Collaborating domestic and international Institutes and Universities on the KSTAR program Domestic Collaborating Domestic Collaborating Universities and Institutes Universities and Institutes IAEA FEC PARK - 5 - Device parameters of KSTAR, ITER and K-DEMO KSTAR KSTAR ITER K-DEMO Parameters (achieved) (Goals) (Baseline/SS) (Option II) Major radius, R0 [m] 1.8 1.8 6.2 6.8 Minor radius, a [m] 0.5 0.5 2.0 2.1 Elongation, 2.16* 2.0 1.7 1.8 Triangularity, 0.8 0.8 0.33 0.63 Plasma shape DN, SN DN, SN SN DN (SN) Plasma current, IP [MA] 1.0 2.0 15/ ~10 > 12 Toroidal field, B0 [T] 3.5 3.5 5.3 7.4 H-mode duration [sec] 73 300 400 / 1000 SS N 3.0 (4.3, transient) 5.0 ~3.0 ~ 4.2 Bootstrap current, fbs ~0.5 ~ 0.6 Superconductor Nb3Sn, NbTi Nb3Sn, NbTi Nb3Sn, NbTi Nb3Sn, NbTi Heating /CD [MW] ~10 ~ 28 ~ 73 120 PFC / Divertor C, W C, W W W Fusion power, Pth [GW] ~0.5 ~ 3.0 * exceeded IAEA FEC PARK - 6 - Topics to be addressed . Mission and role of KSTAR research . Advantages of tokamak plasma research on KSTAR . Progress of the KSTAR research program . High performance and steady-state operation . Physics of MHD/Turbulence . Progress and understanding of the ELM-crash control with RMP . Research efforts for ITER and K-DEMO . Future upgrade plan . Summary IAEA FEC PARK - 7 - Uniqueness-1: nearly perfect symmetry of the tokamak plasmas Extremely low magnetic ripple for minimum ripple loss. (B/B~0.05 %) -5 Extremely low error field configuration (훿B/B0~1x10 ) . Manufacturing and assembly of the magnets with high accuracy . Robust structures to sustain the TF coil shape uniform Designed with strong shaping capability for high beta operation stability . Double & single null capable and passive stabilizer for stabilizing VDE & RWM . Control flexibility using four pairs of CS coils and high elongation ( ~ 2.16) & triangularity ( ~ 0.8) SC magnets manufactured without internal joints to minimize error field IAEA FEC PARK - 8 - Uniqueness-2: Versatile in-vessel control coils for perturbation study and advanced imaging diagnostics Unique in-vessel control coils with 3 poloidal rows KSTAR In-vessel (top, middle and bottom; e.g., ITER) Control Coils Top / Middle / Bottom . Flexible 3D field perturbation with n=1, 2 and mixed . Broad band power supply (~10 kHz) and patch panel for fast control for position, ELM-crash, rotation control & RWM control Advanced 2D/3D imaging diagnostics . ECEI, MIR, BES, imaging bolometer, ECEI MIR Bolometer FILD, SXR, etc. Enable us to access the dynamics of fundamental physics of MHD/ turbulence that were not available IAEA FEC PARK - 9 - Topics to be addressed . Mission and role of KSTAR research . Advantages of tokamak plasma research on KSTAR . Progress of the KSTAR research program . High performance and steady-state operation . Physics of MHD/Turbulence . Progress and understanding of the ELM-crash control with RMP . Research efforts for ITER and K-DEMO . Future upgrade plan . Summary IAEA FEC PARK - 10 - Effect of high toroidal rotation (or high edge rotation shear) on energy confinement in KSTAR Toroidal rotation (Mi or torque) is a hidden parameter in energy Beam torque is linear with confinement scaling?? Mach number . Excellent correlation between the energy confinement time (tth) and Mach number (Mi) or torque . L-mode and H-mode data can be combined with the rotation related parameter (turbulence physics basis?) MAST, Mi ~1. Further study on edge shear, counter rotation, role of intrinsic Off-set value: rotation, etc. is in progress Intrinsic rotation S.G. Lee, EX/P7-4 J.S. Kang, NFRI High toroidal rotation speed in KSTAR may be due to nearly perfect tokamak plasma symmetry (low error field) H-mode L-mode Beam torque is linear with the Mach number (record high Mi~0.8) Edge rotation shear and Er are well correlated with core rotation speed Off-set value: Intrinsic rotation (AU) IAEA FEC PARK - 11 - NBI-heated long pulse H-mode operation (> 1 min) Long pulse H-mode discharge with NBI & ECH • Fully non-inductive advanced operation with high beta (N ~2.0, P ~2.5) was up to 20s. • H-mode was sustained over 70 s, BUT gradual performance degradation due to the heated limiter In 2018 campaign, longer steady-state, fully non- inductive operation over ~100s will be attempted with upgraded - Active water cooling into PFC/Divertor - Boron powder dropper (PPPL / Gradual performance degradation ORNL collaboration) Due to increased wall recycling - Additional heating (NBI, ECCD) 0 10 20 30 40 50 60 70 IAEA FEC PARK - 12 - Exploring conventional and new operation modes for optimal scenarios applicable to ITER and K-DEMO 7 6 5 #16498 4 ITB foot (keV) i 3 T (ρ=0.46) 2 1 0 400 ITB @2.5 s H-mode @4.5 s 300 200 (km/s) t V 100 0 1.8 1.9 2.0 2.1 2.2 2.3 R (m) Search of alternative new mode of Operation (combine ITB and low q95) . Broad ITB provides good confinement in the core . Low q95 avoids harmful MHDs (e.g. NTM and ELMs) . Sawtooth can be used for particle control and q~1 surface can be controlled with CD. Expense for MHD control (ELM and NTM) can be used for additional CD due to reduced Fbs from H-mode IAEA FEC PARK - 13 - Topics to be addressed . Mission and role of KSTAR research . Advantages of tokamak plasma research on KSTAR . Progress of the KSTAR research program . High performance and steady-state operation . Physics of MHD/Turbulence . Progress and understanding of the ELM-crash control with RMP . Research efforts for ITER and K-DEMO . Future upgrade plan . Summary IAEA FEC PARK - 14 - Theoretical model for Sawtooth instability (Full or Partial)? #18186 Direct q0 measurements (2015, 2016) alone cannot validate the model • During nearly four decades, there are two groups of measurement; <q0>~1.0 and <q0>~0.8 with q0 ~0.07 Supplementary experiment is essential • Excitation of double tearing modes inside the q~1 Measurement of q for two years surface is consistent with the growth rates 0 (2015, 2016) confirms <q >~1.0 with calculated by M3DC1 with q >1.0 in MHD 0 0 q0~0.07 (J.
Load more

Recommended publications
  • Engineering Design Evolution of the JT-60SA Project
    EngineeringEngineering DesignDesign EvolutionEvolution ofof thethe JTJT --60SA60SA ProjectProject P. Barabaschi, Y. Kamada, S. Ishida for the JT-60SA Integrated Project Team EU: F4E-CEA-ENEA-CNR/RFX-KIT-CRPP-CIEMAT-SCKCEN JA: JAEA 1 JTJT --60SA60SA ObjectivesObjectives •A combined project of the ITER Satellite Tokamak Program of JA-EU (Broader Approach) and National Centralized Tokamak Program in Japan. •Contribute to the early realization of fusion energy by its exploitation to support the exploitation of ITER and research towards DEMO. ITER DEMO Complement Support ITER ITER towards DEMO JT-60SA 2 TheThe NewNew LoadLoad AssemblyAssembly • JT60-U: Copper Coils (1600 T), Ip=4MA, Vp=80m 3 • JT60-SA: SC Coils (400 T), Ip=5.5MA, Vp=135m 3 JTJTJT-JT ---60U60U JTJTJT-JT ---60SA60SA JT-60SA(A ≥2.5,Ip=5.5 MA) ITER (A=3.1,15 MA) 3.0m 6.2m ~2.5m ~4m 1.8m KSTAR (A=3.6, 2 MA) 1.7m EAST (A=4.25,1 MA) 1.1m 3 SST-1 (A=5.5, 0.22 MA) 3 HighHigh BetaBeta andand LongLong PulsePulse • JT-60SA is a fully superconducting tokamak capable of confining break- max even equivalent class high-temperature deuterium plasmas ( Ip =5.5 MA ) lasting for a duration ( typically 100s ) longer than the timescales characterizing the key plasma processes, such as current diffusion and particle recycling. • JT-60SA should pursue full non-inductive steady-state operations with high βN (> no-wall ideal MHD stability limits). 4 4 PlasmaPlasma ShapingShaping • JT-60SA will explore the plasma configuration optimization for ITER and DEMO with a wide range of the plasma shape including the shape of ITER , with the capability to produce both single and double null configurations.
    [Show full text]
  • Overview of Versatile Experiment Spherical Torus (VEST)
    Overview of Versatile Experiment Spherical Torus (VEST) Y.S. Hwang and VEST team March 29, 2017 CARFRE and CATS Dept. of Nuclear Engineering Seoul National University 23rd IAEA Technical MeetingExperimental on Researchresults and plans of Using VEST Small Fusion Devices, Santiago, Chille Outline Versatile Experiment Spherical Torus (VEST) . Device and discharge status Start-up experiments . Low loop voltage start-up using trapped particle configuration . EC/EBW heating for pre-ionization . DC helicity injection Studies for Advanced Tokamak . Research directions for high-beta and high-bootstrap STs . Preparation of long-pulse ohmic discharges . Preparation for heating and current drive systems . Preparation of profile diagnostics Long-term Research Plans Summary 1/36 VEST device and Machine status VEST device and Machine status 2/36 VEST (Versatile Experiment Spherical Torus) Objectives . Basic research on a compact, high- ST (Spherical Torus) with elongated chamber in partial solenoid configuration . Study on innovative start-up, non-inductive H&CD, high and innovative divertor concept, etc Specifications Present Future Toroidal B Field [T] 0.1 0.3 Major Radius [m] 0.45 0.4 Minor Radius [m] 0.33 0.3 Aspect Ratio >1.36 >1.33 Plasma Current [kA] ~100 kA 300 Elongation ~1.6 ~2 Safety factor, qa ~6 ~5 3/36 History of VEST Discharges • #2946: First plasma (Jan. 2013) • #10508: Hydrogen glow discharge cleaning (Nov. 2014) Ip of ~70 kA with duration of ~10 ms • #14945: Boronization with He GDC (Mar. 2016) Maximum Ip of ~100 kA • # ??:
    [Show full text]
  • Past, Present and Future of the US-KSTAR Collaboration Hyeon K
    Past, Present and Future of the US-KSTAR Collaboration Hyeon K. Park POSTECH at 2009 US-KSTAR Workshop GA.SanDiego, CA On April 15-16, 2009 Past of the international fusion effort • Three large tokamak era: non-steady state device based on Cu coils (pulse length is limited by the cooling system < ~ 20 sec.) – Tokamak Fusion Test Reactor (USA) 1982-1997, Princeton Plasma Physics Laboratory, USA ¾ Fusion power yield: Q ~ 0.3 from D-T experiment – Joint European Tokamak (EU):1983 – present, Culham, Oxfordshore, UK ¾ Fusion power yield: Q ~ 0.7 from D-T experiment – JT-60U (Japan):1985 - present, Japan Atomic Energy Agency (JAEA), Japan ¾ Q~1.25 extrapolated from D-D experiment Internal view of Internal view of TFTR Internal view of JT60-U JET/plasma discharge Future (ITER) • The goal is "to demonstrate the scientific and technological feasibility of fusion power for peaceful purposes". – Demonstration of fusion power yield; Q (output power/input power) ~10 – International consortium (Europe, Japan, USA, Russia, Korea, China, and India) – Total cost ~ $10 B for ~10 years Physics basis is empirical energy confinement scaling KSTAR-US collaboration (past) • US-KSTAR workshop at GA on May 2004 • Areas of interest (first priority of KSTAR)~$1.62M – Steady-state Technology ($14.3M) ($0.37M) – Control, Stability and AT modes($1.48M) ($0.35M) – Conventional Diagnostics ($ 6.5M) ($0.45M) – Advanced Diagnostics ($2.8M) ($0.2M) – Collaboratory ($1.25M)($0.25M) • US-KSTAR workshop 2005 at Daejeon (active work) • US-KSTAR workshop 2006 at Princeton – FY06 International Collaboration to prepare for KSTAR operation (Finals 04/06; $1352K total) – Total allocation for Institution: PPPL (~500k), GA (~400k), ORNL (~200k), LLNL (~20k), MIT(~40k) Columbia (~100k), Escalated KSTAR progress • US-KSTAR workshop (Sept, 2007), Dajeon, Korea – Korean National Assembly passed the Fusion Energy Developm ent Promotion Act on November 30, 2006.
    [Show full text]
  • TH/P8-21 Transport Simulations of KSTAR Advanced Tokamak
    1 TH/P8-21 Transport Simulations of KSTAR Advanced Tokamak Scenarios Yong-Su Na 1), 2), C. E. Kessel 3), J. M. Park 4), J. Y. Kim 1) 1) National Fusion Research Institute, Daejeon, Korea 2) Department of Nuclear Engineering, Seoul National University, Seoul, Korea 3) Princeton Plasma Physics Laboratory, Princeton, NJ, USA 4) Oak Ridge National Laboratory, Oak Ridge, TN, USA e-mail contact of main author: [email protected] Abstract. Predictive modeling of KSTAR operation scenarios are performed with the aim of developing high performance steady state operation scenarios. Various transport codes are employed for this study. Firstly, steady state operation capabilities are investigated with time dependent simulations using a free-boundary transport code. Secondly, reproducibility of high performance steady state operation scenario from an existing tokamak to KSTAR is investigated using the experimental data from other tokamak device. Finally, capability of DEMO-relevant advanced tokamak operation is investigated in KSTAR. From those simulations, it is found that KSTAR is able to establish high performance steady state operation scenarios. The selection of the transport model and the current ramp up scenario is also discussed which have strong influence on target profiles. 1. Introduction As the fusion era is rapidly approaching, the necessity of development of steady state operation scenarios becomes more and more important, particularly for fusion reactor models based on the tokamak concept. In addition to the steady state operation, fusion performance of the tokamak needs to be improved compared with conventional H-modes for developing economically viable fusion power plants. In this context, the, so-called, advanced tokamak (AT) scenarios are being developed aiming at satisfying these two reactor requirements simultaneously.
    [Show full text]
  • Modeling and Control of Plasma Rotation for NSTX Using
    PAPER Related content - Central safety factor and N control on Modeling and control of plasma rotation for NSTX NSTX-U via beam power and plasma boundary shape modification, using TRANSP for closed loop simulations using neoclassical toroidal viscosity and neutral M.D. Boyer, R. Andre, D.A. Gates et al. beam injection - Topical Review M S Chu and M Okabayashi To cite this article: I.R. Goumiri et al 2016 Nucl. Fusion 56 036023 - Rotation and momentum transport in tokamaks and helical systems K. Ida and J.E. Rice View the article online for updates and enhancements. Recent citations - Design and simulation of the snowflake divertor control for NSTX–U P J Vail et al - Real-time capable modeling of neutral beam injection on NSTX-U using neural networks M.D. Boyer et al - Resistive wall mode physics and control challenges in JT-60SA high scenarios L. Pigatto et al This content was downloaded from IP address 128.112.165.144 on 09/09/2019 at 19:26 IOP Nuclear Fusion International Atomic Energy Agency Nuclear Fusion Nucl. Fusion Nucl. Fusion 56 (2016) 036023 (14pp) doi:10.1088/0029-5515/56/3/036023 56 Modeling and control of plasma rotation for 2016 NSTX using neoclassical toroidal viscosity © 2016 IAEA, Vienna and neutral beam injection NUFUAU I.R. Goumiri1, C.W. Rowley1, S.A. Sabbagh2, D.A. Gates3, S.P. Gerhardt3, M.D. Boyer3, R. Andre3, E. Kolemen3 and K. Taira4 036023 1 Department of Mechanical and Aerospace Engineering, Princeton University, Princeton, NJ 08544, USA I.R. Goumiri et al 2 Department of Applied Physics and Applied Mathematics,
    [Show full text]
  • Introduction to Nuclear Fusion
    Introduction to Nuclear Fusion Prof. Dr. Yong-Su Na To build a sun on earth - Open magnetic confinement - Closed magnetic confinement 2 What is closed magnetic confinement? 3 Open Magnetic System B sin 2 min Bmax v|| loss cone loss cone - Suffering from end losses J.P. Freidberg, “Ideal Magneto-Hydro-Dynamics”, lecture note A. A. Harms et al, “Principles of Fusion Energy”, World Scientific (2000) 4 Open Magnetic System Magnetic field Is this motion realistic? ion Dunkin donuts (2010) 5 Closed Magnetic System Magnetic field Donut-shaped vacuum vessel ion 6 Closed Magnetic System 7 Closed Magnetic System Magnetic field R 0 a Plasma needs to be confined ion R0 = 1.8 m, a = 0.5 m in KSTAR 8 Closed Magnetic System Magnetic field R 0 a Plasma needs to be confined ion R0 = 6.2 m, a = 2.0 m in ITER 9 Closed Magnetic System Toroidal Field (TF) coil Magnetic field Toroidal direction Applying toroidal magnetic field ion 3.5 T in KSTAR, 5.3 T in ITER 10 Closed Magnetic System Toroidal Field (TF) coil Toroidal direction Applying toroidal magnetic field 3.5 T in KSTAR, 5.3 T in ITER 11 Closed Magnetic System Toroidal Field (TF) coil Magnetic field Toroidal direction Magnetic field of earth? 0.5 Gauss = 0.00005 T ion 12 http://www.crystalinks.com/earthsmagneticfield.html Closed Magnetic System Magnetic field Magnetic field of earth? 0.5 Gauss = 0.00005 T ion http://www.transformacionconciencia.com/archives/2384 13 Closed Magnetic System Magnetic field ion 14 Lesch, Astrophysics, IPP Summer School (2008) Closed Magnetic System Magnetic field ion electron
    [Show full text]
  • Nuclear Fusion Research Activities at KAERI
    Transactions of the Korean Nuclear Society Spring Meeting Jeju, Korea, May 12-13, 2016 Nuclear Fusion Research Activities at KAERI Dong Won Leea, Sun Ho Kima, Seong Ho Jeonga, Byung Hoon Oha aKorea Atomic Energy Research Institute, Republic of Korea *Corresponding author: [email protected] 1. Introduction Large aspect ratio (5.6) High bootstrap current (≥50%) Nuclear fusion is considered to be a next generation Intensive RF heating and current drive (5 clean and sustainable energy due to its inherent safety MW) and abundant fuel resource. In this context, ITER has been built to resolve the scientific and technological issues remained for the ignition at Cadarache in France 3. Research & Development on KSTAR heating since 2006 [1, 2]. Korea has joined the ITER project system and contributed to ITER construction and understanding of plasma physics through KSTAR We have participated in KSTAR construction and (Korea Super Conducting Tokamak Advanced operation using the lessons from the former tokamaks Research) [3]. [4-6] with focus on tokamak heating and current drive KAERI has much experience on fusion plasmas devices such as ICRF and NB since 1996. It provided through KT-1 development and KT-2 planning since KSTAR heating power up to 6MW at present time. 1983. After that we have participated the KSTAR and Especially, NB contributed to achievement of long- ITER projects in various fields. In the present paper, pulse stable H-mode during 40 sec at KSTAR [7]. It these activities at KAERI, especially for Fusion Nuclear will enable KSTAR to achieve high beta long pulse Engineering Development Division were introduced.
    [Show full text]
  • The Advanced Tokamak: Goals, Prospects and Research Opportunities Amanda Hubbard MIT Plasma Science and Fusion Center with Thanks to Many Contributors, Including A
    The Advanced Tokamak: Goals, prospects and research opportunities Amanda Hubbard MIT Plasma Science and Fusion Center with thanks to many contributors, including A. Garafolo, C. Greenfield, C. Kessel, D. Meade, M. Murakami, F. Najmabadi, T. Taylor Opinions are my own… GCEP Fusion Energy Workshop on Opportunities for Fundamental Research and Breakthrough in Nuclear Fusion Princeton, NJ May 1-2 2006 The Advanced Tokamak • Introduction: What is an ‘advanced tokamak’? • The AT vision for fusion energy – Drawing heavily on ARIES studies. • Current results and near-term prospects – Focusing here on US program. • AT on ITER: What we will (and won’t) learn. • Research Opportunities: ideas to advance and accelerate fusion energy prospects. To start the discussion…. An “advanced tokamak” device is, in terms of magnetic configuration, simply a TOKAMAK Pure toroidal field does not confine charged particles Tokamak needs a toroidal current for stability. Adding poloidal field does Current conventionally driven by tranformer; confine charged particles. - Current is driven around central solenoid. Produced by toroidal current. Inherently NOT steady-state. Tokamaks lead other configurations in fusion performance, are approaching ‘breakeven’ ITER D. Meade, ARIES workshop 4/24/05 “Conventional” tokamak operation will be primary mode of operation on ITER • Heating applied mainly on-axis, inductive current drive, profiles relax to ‘natural’ state. • Much experience worldwide, good confidence in extrapolation to burning plasma conditions. – This will allow
    [Show full text]
  • Poster Session 1 Author Title Session Nr
    Poster Session 1 Author Title Session Nr. Chen, Y. Design of Real-time Data Acquisition System for POlarimeter- 1,3 INTerferometer diagnostic Fuchs, J.C. CLISTE equilibrium reconstruction in ASDEX Upgrade using 1 toroidally distributed magnetic measurements and possible implications on the plasma position control Gerkšič, S. SVD-based fast online MPC for ITER plasma current and shape 1 control Guo, Y. Preliminary results of a new MIMO plasma shape controller for 1 EAST Huang, Y. Improvement of GPU parallel real-time equilibrium reconstruction 1 for plasma control Li, C.C. Preliminary Implementation of the Real-time Data Sharing System 1,7 Based on RFM for EAST Manduchi, G. A Portable Control And Data Acquisition Solution 1 Shah, M. FPGA based Trigger and Timing System for Aditya-U Tokamak 1 Operation and Control Shin, G.W. Automatic Detection of L-H Transition in KSTAR by Support Vector 1 Machine Signoret, J. Coupling Plasma Control System and Pulse Schedule Editor for 1 WEST Operation StStepanov, DD. ITER CODAC OOperation ti RRequest t Gatekeeper: G t k First Fi t 1 Implementation Varshney, S. Design maturity of Plant Instrumentation and Control for ITER 1 XRCS-Survey Vega, J. Real-time implementation with FPGA-based DAQ system of a 1 probabilistic disruption predictor from scratch Wang, Y. Closed-loop control-oriented system identification for EAST 1 plasma shape response model Wu, K. Radiative Feedback for Divertor Heat Spread on EAST 1 Xiao, B.J. Model-based plasma vertical stabilization and position control at 1 EAST Yeole, Y.G. An Automation of Vacuum System for SMARTEX-C Yuan, Q.P.
    [Show full text]
  • Nuclear Fusion Power – an Overview of History, Present and Future
    International Journal of Advanced Network, Monitoring and Controls Volume 04, No.04, 2019 Nuclear Fusion Power – An Overview of History, Present and Future Stewart C. Prager Department of Physics University of Wisconsin – Madison Madison, WI 53706, USA E-mail: [email protected] Summary—Fusion power offers the prospect of an allowing the nuclei to fuse together. Such conditions almost inexhaustible source of energy for future can occur when the temperature increases, causing the generations, but it also presents so far insurmountable ions to move faster and eventually reach speeds high engineering challenges. The fundamental challenge is to enough to bring the ions close enough together. The achieve a rate of heat emitted by a fusion plasma that nuclei can then fuse, causing a release of energy. exceeds the rate of energy injected into the plasma. The main hope is centered on tokamak reactors and II. FUSION TECHNOLOGY stellarators which confine deuterium-tritium plasma In the Sun, massive gravitational forces create the magnetically. right conditions for fusion, but on Earth they are much Keywords-Fusion Energy; Hydrogen Power; Nuclear Fusion harder to achieve. Fusion fuel – different isotopes of hydrogen – must be heated to extreme temperatures of I. INTRODUCTION the order of 50 million degrees Celsius, and must be Today, many countries take part in fusion research kept stable under intense pressure, hence dense enough to some extent, led by the European Union, the USA, and confined for long enough to allow the nuclei to Russia and Japan, with vigorous programs also fuse. The aim of the controlled fusion research underway in China, Brazil, Canada, and Korea.
    [Show full text]
  • Lessons Learned from KSTAR Construction and Commissioning
    J. Plasma Fusion Res. SERIES, Vol. 9 (2010) Lessons Learned from KSTAR Construction and Commissioning J. S. Bak, H. L. Yang, Y. K. Oh, Y. S. Kim, Y. M. Park, K. R. Park, W. C. Kim, M. K. Park, Y. S. Bae, I. S. Whang, K. W. Cho, Y. J. Kim, M. Kwon, G. S. Lee and the KSTAR Team National Fusion Research Institute (Received: 15 October 2009 / Accepted: 9 March 2010) KSTAR is an advanced tokamak with fully superconducting coils for the steady state research, which began in December 1995 and was completed in July 2008. As a single science project, it has been marked as the largest construction project in the history of Korea. During the construction period, we encountered several challenges, such as technical issues, cost overrun and schedule slippage. Nevertheless, we managed to overcome the difficulties through the devotion of all the participants under a strong leadership, and we finally succeeded in the first plasma discharge of 133kA/865ms in July 2008. Therefore, KSTAR, the first Nb3Sn based fully superconducting tokamak, has been recorded as the device that passed its commissioning without failure at its first trial. And as we speak, KSTAR is sustaining its momentum for the next campaign: it is getting ready for its operation of the toroidal magnetic field of 3 Tesla. At this presentation, we focus on the actual applications and resolutions applied to surmount the difficulties we faced during the KSTAR construction, as well as the events that occurred behind the scenes. Despite the scale difference between KSTAR and ITER, we strongly believe that the presentation will provide a great insight to the construction of the ITER device, which is very similar to KSTAR in terms of design and characteristics.
    [Show full text]
  • Design and Construction of the KSTAR Tokamak
    IAEA-CN-77/OV7/1 Design and Construction of the KSTAR Tokamak G. S. Lee and the KSTAR Team National Fusion R&D Center, Korea Basic Science Institute, Taejon, Republic of Korea e-mail contact of main author: [email protected] Abstract. The extensive design effort has been focused on two major aspects of the KSTAR project mission, steady-state operation capability and “advanced tokamak” physics. The steady-state aspect of mission is reflected in the choice of superconducting magnets, provision of actively cooled in-vessel components, and long- pulse current-drive and heating systems. The “advanced tokamak” aspect of the mission is incorporated in the design features associated with flexible plasma shaping, double-null divertor and passive stabilizers, internal control coils , and a comprehensive set of diagnostics. Substantial progress in engineering has been made on superconducting magnets, vacuum vessel, plasma facing components, and power supplies. The new KSTAR experimental facility with cryogenic system and de-ionized water-cooling and main power systems has been designed, and the construction work has been on-going for completion in year 2004. 1. Introduction The mission of the Korea Superconducting Tokamak Advanced Research (KSTAR) Project is to develop a steady-state-capable advanced superconducting tokamak to establish a scientific and technological basis for an attractive fusion reactor [1,2]. To support this project mission, three major research objectives have been established: (i) to extend present stability and performance boundaries of tokamak operation through active control of profiles and transport, (ii) to explore methods to achieve steady-state operation for tokamak fusion reactors using non-inductive current drive, and (iii) to integrate optimized plasma performance and continuous operation as a step toward an attractive tokamak fusion reactor.
    [Show full text]